1. Field of the Invention
The present invention relates generally to heat exchangers, and more particularly to a new and improved aperture, through-bore, or hole design or configuration which is to be defined within each tube support plate of a liquid metal-cooled fast breeder nuclear reactor (LMFBR) steam generator heat exchanger so as to be capable of supporting each of the steam generator heat exchange tubes relative to each of the tube support plates in such a manner that deleterious wear effects upon each of the heat exchange tubes, as a result of relative movement defined between each one of the heat exchange tubes and any one of the tube support plates, is minimized.
2. Description of the Prior Art
A nuclear reactor produces heat as a result of the fission of nuclear material which is disposed within fuel rods, the fuel rods being secured together so as to define fuel assemblies. The fuel assemblies define the nuclear reactor core, and the core is disposed within a reactor or pressure vessel. In commercial nuclear reactor facilities, the heat produced by means of the fission processes is utilized to generate electricity. In particular, conventional facilities usually comprise one or more primary flow and heat transfer or exchange loops, and a corresponding number of secondary flow and heat transfer or exchange loops to which conventional steam generators and steam turbines, as well as electrical generators, are fluidically and mechanically connected, respectively. A typical energy conversion process for such commercial nuclear reactor facilities would therefore comprise the transfer of heat from the nuclear core to the primary coolant flow and loop system, from the primary coolant flow and loop system to the secondary coolant flow and loop system by means of suitable heat exchangers, and finally from the secondary coolant flow and loop system to the steam generators by means of further suitable heat exchangers. The generated steam is then of course transmitted to the steam turbines to which the electrical generators are operatively connected, and from which electricity is ultimately generated.
In a liquid metal-cooled fast breeder reactor, liquid sodium serves as the reactor coolant and is therefore circulated through the primary coolant flow and loop system which typically comprises the nuclear core, a heat exchanger, and a circulating pump. In nuclear reactors having more than one primary coolant flow loop within the primary coolant flow system, the nuclear core and the reactor pressure vessel, within which the nuclear core is disposed, are connected in common to each of the primary coolant flow loops. The heat generated by means of the nuclear core is thus removed therefrom by means of the reactor coolant which is conducted into the reactor vessel and through the reactor core. The heated reactor coolant then exits from the nuclear core and the reactor vessel so as to flow through the heat exchangers which serve to transfer the heat to the secondary flow system loops operatively associated therewith. Liquid sodium is likewise disposed within the secondary flow system loops, and, in turn, the heated sodium disposed within the secondary flow system loops passes through suitable heat exchangers so as to transfer its heat to the water being conducted through the steam generators whereby steam is generated for use within the steam turbines. The steam turbines are then of course utilized to drive the electric generators for generating electricity. The cooled reactor core sodium coolant disposed within the primary flow loop system is of course recirculated back to the reactor pressure vessel and the reactor core by means of the primary flow loop system circulating pump, the secondary sodium coolant is likewise recirculated by means of its recirculating pump, and the coolant cycles are repeated. An intermediate or secondary coolant loop system is acknowledged to be mandatory within a liquid metal-cooled fast breeder reactor in view of the fact that liquid sodium exhibits explosive instability in the presence of water. Consequently, should any leakage develop between the sodium and water flow loops or paths, it is imperative to, in effect, isolate the reactor core from any deleterious effects attendant the violent or explosive mixing of the sodium and water components. In addition, the sodium disposed within the primary coolant flow loop system becomes radioactive after having been conducted through the reactor core, whereas the sodium coolant disposed within the closed secondary flow loop system is not radioactive. Consequently, should any leakage develop within the primary coolant loop system, the secondary coolant loop system effectively serves as a buffer zone to maintain radioactive coolant out of the steam generator water and steam, and prevent contamination of the steam turbines and the electric generator equipment.
Within one type of conventional, exemplary heat exchanger system defined between the intermediate or secondary non-radioactive liquid sodium coolant loop and the steam generator tubes through which water is conducted for the generation of steam, the water-steam generator tubes extend substantially vertically within the heat exchanger with the upper and lower ends of the tubes fixedly secured within upper or top, and lower or bottom, tubesheets, respectively. Suitable headers or manifolds fluidically surround the ends of the tubes and encase the tubesheets so as to supply water to, for example, the lower ends of the tubes and extract steam from the upper ends of the tubes. The liquid sodium enters the upper end of the heat exchanger at an elevational level below the upper or top tubesheet, and relatively cooled liquid sodium leaves the heat exchanger at an elevational level above the lower tubesheet. Consequently, the initially hot liquid sodium and initially cold water flow in opposite directions through the heat exchanger whereby the heat exchange process is conducted therebetween. In order to substantially restrain excessive lateral movements of the water-steam tubes within the heat exchanger under the influence of the downwardly flowing liquid sodium, which may exhibit cross-current flow paths or the like, or to substantially confine lateral movements of the water-steam tubes under any vibrational conditions which may manifest themselves, a plurality of vertically spaced tube support plates are fixedly suspended within the heat exchanger at positions intermediate the upper and lower tubesheets. The tube support plates are suspended in a stacked array from the upper or top tubesheet by means of a plurality of stay rods interposed between, and fixedly secured to, each tube support plate, and the entire array freely hangs downwardly from the upper or top tubesheet and is not fixed to the lower or bottom tubesheet. The water-steam tubes therefore pass through suitable apertures defined within the tube support plates such that the tubes are not fixedly secured to the tube support plates. Additional apertures are also provided within the tube support plates so as to permit the liquid sodium to pass therethrough, the liquid sodium also passing through the annular spaces defined between the water-steam tubes and the sidewalls of the apertures of the tube support plates through which the water-steam tubes pass, whereby the liquid sodium can traverse the vertical extent of the heat exchanger in furtherance of performance of its heat exchange function.
Within the aforenoted type of heat exchanger system, an operational deficiency has developed to the effect that excessive wear, and therefore a premature service life, of the water-steam tubes has manifested itself. The reason that such unacceptably excessive wear problems develop is due to the fact that during normal reactor operation, there is considerable movement of the water-steam tubes relative to the tube support plates, as well as considerable movement of the tube support plates relative to the water-steam tubes. For example, during start-up or transient operational conditions, the water-steam tubes may experience substantial vibrational loads which may cause movement of the same relative to the tube support plates. Cross-currents of the liquid sodium flowing through the heat exchanger may likewise cause transverse movement or buckling loads to be impressed upon the water-steam tubes thereby likewise causing movement of the same relative to the tube support plates. Such relative transverse movement of the water-steam tubes with respect to the tube support plates causes abrasive wear to be developed between the water-steam tubes and the sidewalls of the tube support plate apertures through which the water-steam tubes pass. Alternatively, during the start-up period of the power plant facility, the tube support plates and their stay rods are relatively cool and are at the same temperature as the water-steam tubes. However, as the plant or facility becomes operational, the tube support plates and their stay rods become very hot due to the fact that the heated liquid sodium is in direct contact therewith throughout the heat exchanger. To the contrary, the water-steam tubes will not experience a similar elevation in temperature level due to the fact that cold water is being initially circulated within the tubes and a change of state from water to steam absorbs a substantial amount of the heat energy through means of the heat exchange process. Consequently, the tube support plates, and particularly the stay rods thereof supporting the same, will experience substantial thermal growth whereby the tube support plates will in fact exhibit considerable movement relative to the water-steam tubes. In a similar manner, transient operating conditions may also result in variable thermal conditions prevailing within the heat exchanger whereby thermal expansion or contraction of the tube support plates and their stay rods may again be manifested. For example, should a pump failure occur within the liquid sodium coolant loop circulating system, the heated liquid sodium will be stagnantly disposed within the loop and the heat exchanger while the relatively cooler water will continue to be conducted through the water-steam tubes. A relative increase in the temperature level of the sodium will therefore be experienced relative to the water-steam tubes, and consequently, it follows that the tube support plates and the stay rods will likewise experience an elevation in temperature and thermal growth. Upon rectification of the circulating pump malfunction, steady-state operations will again prevail whereby the temperature level of the liquid sodium will be somewhat lowered whereby the tube support plates and stay rods would experience some thermal contraction relative to the water-steam tubes. All of such relative movements result in unacceptably excessive wear of the water-steam tubes whereby the same have to be replaced more frequently than would normally be expected or desired.
In order to alleviate the foregoing problems, it has been attempted to re-design the sidewalls of the tube support plate apertures through which the water-steam tubes pass so as to minimize the aforenoted excessive wear problems. In lieu of conventional, straight through-bores defined within the tube support plates whereby when the water-steam tubes become angularly oriented with respect to the tube support plates due to the aforenoted relative movements of either the tubes or the tube support plates with respect to each other, or simply when the tubes are in contact with the tube support plate aperture sidewalls due to eccentric alignment defined therebetween, excessive wear of the tubes was experienced due to abrasion developed between the tubes and the upper or lower annular edges of the tube apertures defined within the tube support plates, one attempt sought to provide such upper and lower aperture annular edges with rounded or chamfered edge surfaces. In accordance with another prior art attempt at resolving the aforenoted abrasion problems, the upper and lower regions of the through-bores or apertures were actually provided with conically-configured tapered surfaces which met at a central land portion. As may be surmised, however, despite both attempts at such a resolution of the problem, the problem persisted in view of the fact that the tubes nevertheless had to ride over or traverse the central land region which, in itself, presented a substantial edge portion about which the tubes would be bent and abraded. A still further, third type of attempt to resolve the aforenoted problem consisted in providing the interior sidewall portions of the tube support plate apertures with radially inwardly extending cusp-shaped projections upon which the tubes may be supported either by means of lineal or point contact surfaces. Nevertheless, such an attempted resolution to the problem also failed to in fact resolve the problem, for again, sharp-edged surfaces were present which still generated a considerable amount of abrasion upon the exterior sidewalls of the water-steam tubes.
Accordingly, it is an object of the present invention to provide a new and improved aperture, through-bore, or hole design or configuration within a tube support plate of a heat exchanger through which a water-steam tube of a steam generator is adapted to be disposed.
Another object of the present invention is to provide a new and improved aperture, through-bore, or hole design or configuration within a tube support plate of a nuclear reactor steam generator heat exchanger through which a water-steam tube is adapted to be disposed.
Yet another object of the present invention is to provide a new and improved aperture, through-bore, or hole design or configuration within a tube support plate of a nuclear reactor steam generator heat exchanger through which a water-steam tube is adapted to be disposed, wherein the various operational disadvantages and drawbacks of conventional tube support plate apertures or through-bores will be overcome.
Still another object of the present invention is to provide a new and improved aperture, through-bore, or hole design or configuration within a tube support plate of a nuclear reactor steam generator heat exchanger through which a water-steam tube is adapted to be disposed, wherein unacceptably excessive wear upon the water-steam tube due to relative movement defined between the water-steam tube and the tube support plate will be effectively eliminated or minimized.
Yet still another object of the present invention is to provide a new and improved aperture, through-bore, or hole design or configuration within a tube support plate of a nuclear reactor steam generator heat exchanger through which a water-steam tube is adapted to be disposed, wherein the particular design or configuration of the aperture, through-bore, or hole is such as to readily lend itself to substantially simplified production or fabrication.
Still yet another object of the present invention is to provide a new and improved aperture, through-bore, or hole design or configuration within a tube support plate of a nuclear reactor steam generator heat exchanger through which a water-steam tube is adapted to be disposed, wherein such aperture, through-bore, or hole design or configuration serves to substantially extend the service life of the water-steam tubes within the heat exchanger.